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Title: Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4

Abstract

Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized-water-reactor steam-generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models being developed to provide better digital codes to compute the bahavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

Authors:
 [1]
  1. ed.
Publication Date:
Research Org.:
Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
OSTI Identifier:
6320795
Report Number(s):
NUREG/CR-2716-Vol.4; PNL-4275-4
ON: DE83012248
DOE Contract Number:  
AC06-76RL01830
Resource Type:
Technical Report
Resource Relation:
Other Information: Portions are illegible in microfiche products
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; REACTOR SAFETY; GRAPHITE; COMPRESSION STRENGTH; OXIDATION; HTGR TYPE REACTORS; PWR TYPE REACTORS; RESEARCH PROGRAMS; STAINLESS STEELS; FRACTURE PROPERTIES; STEEL-ASTM-A508; STRESS CORROSION; STEEL-ASTM-A533-B; COMPUTER CALCULATIONS; DECONTAMINATION; EDDY CURRENT TESTING; FATIGUE; FUEL RODS; FUEL-CLADDING INTERACTIONS; LOSS OF COOLANT; PIPES; PRESSURE GRADIENTS; PRESSURE VESSELS; REACTOR CORE DISRUPTION; REACTOR MATERIALS; STEAM GENERATORS; SUPPORTS; ACCIDENTS; ALLOYS; BOILERS; CARBON; CARBON STEELS; CHEMICAL REACTIONS; CHROMIUM ALLOYS; CLEANING; CONTAINERS; CORROSION; CORROSION RESISTANT ALLOYS; ELECTROMAGNETIC TESTING; ELEMENTAL MINERALS; ELEMENTS; FUEL ELEMENTS; GAS COOLED REACTORS; GRAPHITE MODERATED REACTORS; IRON ALLOYS; IRON BASE ALLOYS; MATERIALS; MATERIALS TESTING; MECHANICAL PROPERTIES; MECHANICAL STRUCTURES; MINERALS; NONDESTRUCTIVE TESTING; NONMETALS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SAFETY; STEELS; TESTING; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS; 220900* - Nuclear Reactor Technology- Reactor Safety; 210300 - Power Reactors, Nonbreeding, Graphite Moderated; 210100 - Power Reactors, Nonbreeding, Light-Water Moderated, Boiling Water Cooled; 210200 - Power Reactors, Nonbreeding, Light-Water Moderated, Nonboiling Water Cooled

Citation Formats

Edler, S K. Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4. United States: N. p., 1983. Web. doi:10.2172/6320795.
Edler, S K. Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4. United States. https://doi.org/10.2172/6320795
Edler, S K. 1983. "Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4". United States. https://doi.org/10.2172/6320795. https://www.osti.gov/servlets/purl/6320795.
@article{osti_6320795,
title = {Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4},
author = {Edler, S K},
abstractNote = {Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized-water-reactor steam-generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models being developed to provide better digital codes to compute the bahavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.},
doi = {10.2172/6320795},
url = {https://www.osti.gov/biblio/6320795}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Apr 01 00:00:00 EST 1983},
month = {Fri Apr 01 00:00:00 EST 1983}
}